ML18058B920

Responds to Deviations Noted in Insp Rept 50-255/93-10. Corrective Actions:Eop Guidance for Operators Encl

ML18058B920
Person / Time
Site:	Palisades
Issue date:	07/06/1993
From:	Slade G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9307150102
Download: ML18058B920 (22)

Text

50-255 CPC PATJ1SADF.S INSPECTION REPOR'rS 92-13 '!'HRU q}._.,,. /0

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consumers Power.

l'OWERINli NllCHlliAN'S l'RDliRESS Palisades Nuclear Plant:

27780 Blue Star Memorial Highway, Covert, Ml 49043 July 6, _ 1993

• NutJear Regulatory Commission Document Control. Desk Washington, DC 20555 GB Slade

• General Manager.

DOCKET 50-255 - LICENSE DPR PALISADES PLANT - RESPONSE TO INSPECTION REPORT No. 91010; NOTICE OF DEVIATION AND OPEN ITEMS Inspection report No. 93010 requested a written response to the enclosed Notice*.

of Deviation within 30 days, and a response to the listed open items within 60 days.

On June 18, 1993, a telephone request to combine both responses and submit them by July 7, 1993, was granted by Mr. T Burdick of NRC Region III.

This letter contains that combined response.

• We have reviewed the Notice of Deviation, the open items, and the associated documents.

As a result of our review we agree that our actions; taken to comply with commitments made in our April 28, 1986 letter, do not precisely match those commitments, and that our actions did not fully address the NRC *concerns document in their SER; as noted in open item 93010-03.

However, We do not agree that the deviation stated in the Notice of Deviation occurred.

Detailed di~cussion df the Notice of Deviation is presented in Attachments 1 and 2~ of

• the Open Items, in Attachment 3; and of related commitments, in Attachment 4.

The Inspection Report deals with tha traini~g,.Emergency Operating Procedure (EOP) guidance, and operator actions necessary to compensate for the possibility of a steam line break, inside the containment, occurring concurrently with the failure of a Main Steam Isolation Valve (MSIV) to close.

As stated in the

. inspection report, this event could result in the blowdown of both steam generators into the containment, with temperature and pressure exceeding the containment design and the electrical equipment qualification envelope.

This occurrence will be referred to hereinafter as the "event 11 The inspection report precedes its discussion of subjects involving the "event" with a brief synopsis which is based on the transient analyses supporting the PRA and on simulator runs.

Si~ilarly, we will precede our reply* with a discussion of that synopsis.

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2 The sequence of events listed in the synopsis does generally represent what could happen for the "event", with a large break size. There are some details which w~ believe are not entirely correct or consistent.

1)

• The transient analyses which supports the PRA for the "event" contain graphs for several different case~. - Many of these cases were computer

• runs, or results of other_ analyses,- included to provide answers to specific questions from the staff. The data should only be used in the context of the related question. Several runs compare particular parameter responses for single and double Steam Generator (S/G) blowdowns.

The graphs following NRC question 3.f, regarding the containment re~ponse to differing break sizes, are for single S/G blowdowns.

Mixing data from different cases can lead to incorrect conclusions.

2)

The synop~is makes the following statement: *

3)

Upon dry out of the "Most Affected" steam generator the PCS begins to reheat and repressurize. The saturation temperature in the "Least Affected" steam generator is initia17y above PCS temperature preventing it from acting as a heat sink and inhibiting natural circulation.

The statement is true, momentarily, for certain break sizes.* However, its use in the synopsis implies that there is something unsafe or undesirable about that.

We do not consider the possibility that the "Least Affected S/G may initially be slightly above Primary Coolant System (PCS) temperature at the

• termination of the b 1 owdown to present a core coo 1 i ng prob 1 em.

The word "initially" is important.

During the minutes following these initially occurring conditions, the PCS would reh~at (decay heat additi~n) and the S/G would continue to co_ol (automatic Auxiliary Feedwater initiation).

As sbon as the PCS temperature exceeded the S/G temperature, normal heat transfer and natural circulation would commence; The amount that the saturation temperature bf the "Least Affected" S/G re~ains above the "Most Affected" S/G and the PCS is a function of the differing steam flow from the two S/Gs. _ For example, the case where this te_mperature difference is maximum (the inhibition to initiation of. natural circulation is maximized) is the case of a "normal" Main Steam Line Break (MSLB), i.e. one where the MSIVs both close as designed.

In that case it is not considered a problem to rely o*n the_ "Least Affected" S/G for subsequent core and PCS cooling.

The conclusion that some voids might be formed in-the upper head region is correct for certain situations, but analytical assessments of the situation indicate that void formation in the upper head region would not inhibft natural circulation.

. Both the analyses submitted with the PRA and the preliminary analyses, wh-ich we are now conducting, for the current plant configuration conclude that there is cohsiderable time following the blowdown phase of the "event" for initiation of cooling, whether Auxiliary Feedwater or Once-Through-Cooling.

Analyses done for the Loss of All Feedwater event conclude that successful removal _of decay heat can be accomplished by either Once-Through-Cooling using a single PORV, two charging pumps, and a single High Pressure Safety Injection (HPSI)/Containment Spray pump train, or Auxiliary Feedwater flow using a single pump~ even if i ni ti ated after steam generator dryout, with the* primary_ cool ant temperature

3 approaching 545°F.

The double steam generator blowdown event initially cools the PCS, providing time during the PCS reheating for initiation of either cooling method.

The-availability of the additional train of equipment for each cooling method, which would be available following the subject event, leng~hens the time available for initiation of cooling.

We consider, therefore, that upon completion of the actions ~iscussed in, that Palisades will have not only met our commitments with respect to this issue, but.will have satisfied the NRC concerns st~ted in the SER.

Gerald B Slade General Manager CC Administrator, ReQion III~ USNRC NRC Resident Inspector - Palisades_

Attachments

~.

ATTACHMENT 1 Consumers Power Company Palisades Plant Docket 50-255 Discussion of Notice of Deviation July 6, i993 4 Pages

Attathment 1, Page 1

~otice of Deviation The discussion supporting the Notice of Deviation, contained in the body of the inspection report, implies the NRC concern is: if the subject "event" should occur, the Palisades EOPs do not direct the operators to initiate once-through-=

cooling quickly eriough to assure that the degraded containment environment doei not disable vital equipment prior to its completing its safety f~nction. The Notice of Deviation, however, deals solely with the contention that the EOPs are not structured to accomplish those actions stated* i~ our April 28, 1986 letter.

We do not believe that contention is correct.*

Notice of Deviation Paragraph 2:

In an April 28, 1986 letter to the NRC, the licensee committed, in part, tq perform certain actions for a main steam line.breijk coincident with failure af the intact stea~ gerlerato~'s main steam isolation valve to close.

These specific actions include opening the*

power-operated relief valves, maximizing safety injection flow (feed and bleed), and maximizing auxiliary feed flow to one s"team generator if the decision is made not to rely on any in-containment instrumentation.

CPCo Response:

Palisades agrees with this paragraph.

The Notice of Deviation correctly stat~s that Palisades committed to perform certain actions,if the "event" should occur and if the decision is made not to rely on an*y in-con.tainment instrumentation.

As stated in our April 28, 1986 letter:.

The new EOPs are structured such that, if in-containment vital instrumentation is considered lost then the Functional Recovery Procedure is implemented.

If a decision is made to not rely on any in-containment instrumentation, the result is that 1) the PORVs are opened and a77 available safety injection-(SI) pumps started and SI valves open; 2) auxiliary feed flow is maximized and directed to at least one SIG; 3).containment air coolers are placed in emergency configuration and containment spray manually maximized;

~nd 4) all available service water and component cooling water pumps operated:

• .Attachment 2 provides a listing of the EOP steps which direct the operators to take the stated actions, if the "event" should occur~

We do note that the action "auxiliary feed flow is maximized and directed to at least one S/G" is not implemented exactly.

As noted in Attachment 2, auxiliary feedwater is directed to at least one steam generator, but rather thari directing that AFW flow should be maximized, it is directed that flow be at least 165 gpm, with a footnote stating that as much as 310 gpm may be necessary to recover S/G l~vel.

We did not infer, from the wording of the inspection report, that this difference wa~ the concern which generated the Notice of Deviation.

• • • ~ Page 2 Noti~e of Deviation Paragraph 3:

In the basis document for emergency operating procedures (EOP) 9. 0, "Functional Recovery Procedure," and EOP 6.0, '"Excessive Steam Demand Event," the licensee made the conservative assumptfon that no in-containmerit instrumentation would remain operable.

CPCo'Response:

The Notice of Deviation incbrrectly states "In the basis document fo~ emergency procedures (EOP) g;o, 'Functio'nal Recovery Procedure,' and EOP 6.0, 'Excessive

. Stea~ Demand Event,' the licensee made the conservative assumption that no.in-containment instrumentation would remain operable." The NRC interpretation of the subject basis paragraphs that they co~stitute a deci~ion to ignore all in~

containment i~strumentation if the "event" occurs,* is incorrect.

No advance decision*to ignore in-tbntainment instrumentation has been made.

The subject paragraphs, found in the bases for both EOP 6.0 and 9.0, are similar. Those in EOP 6.0 Basis read as follows:

Th* following references have told the NRC that a major concern of th~

single MSIV failure/two SIG steam line break inside containment issue is the containment temperature and pressure exceeds the EEQ envelope for a77 in-containment instrumentation (and hence exceeds containment design temperature and pressure: Letter dated May 23, 1985 from BD Johnson (CPCo) to Director (NRR) transmitting "Assessment of Palisades Main Steam Line Break.(MSLB) Single Failure Backfits."

This reference in providing a strategy for this specific scenario, make the conservative assumption that no in-containm~nt instrumentation remains operable.

Hence, when design containment *temperatures and pressure are

. exceeded, the Functional Recovery Procedure is implemented.

Th~ intent of the subject paragra~hs of the EOPbases is to simply note that a bounding assumption was made in the PRA analysis of the "event". *That assumption was not intended as either a prediction of expected events, or as.a basis for a decision for the operators to ignore their instrumentation. The assumption was intended as a bounding condition made, for purposes of the PRA analyses only, due to the difficulty of quantifying the failure rate of instrumentation subjected to £onditions beyond those for which it was qualified.

Similar bounding assumptions are made in other s~fety analyses, sometimes (when the bounding case provides acceptable results) to ~void unnecessary analyies, and sometimes to comply with regulatory guidance.

As examples, we assume, in.

the LOCA analysis, that the control rods do not insert and that the reactor is shutdown by the voids,_yet we still direct the operators to make every effort to

• assure the rods are inserted; we assume, in most analyses, that offsite power is*

lost, yet we do not direct the operator to rely solely upon the emergency diesel generators.

• , Page 3

• Noti~e of Deviation Paragraph 4:

Contrary to the above, as of May 4, 1993, the emergency operating procedures do not,decisively instruct the operators to perform the

. above actions~ Instead, EDP 9.0 ambiguously directs the operators to monitor in-containment parameters, which may be inoperable or unreliable, to determine the proper mitigating procedure.

CPCo Response:

Since the EOPs do result in those actions listed in our commitment of April 28, 1986, and since the decision regarding reliance upon instrumentation has been, and must bei left.to the operators on the scene, we believe that, with the minor exception noted regarding auxiliary feedwater, we have fulfilled.the referenced.

commitments.

The Notice of Deviation ~lludes to the EOPs providing ambiguous direction and to their not being dec.isive.

These are subjective judgements which we do not share.

We consider that the Palisades EOPs, when combined with the associated*

• operator traini~g, provide definite guidance for the operatori*t9 accomplish the desired tasks.

, Page 4 NRC Inspection Report Discussion Associated with Notice of Deviation:

The licensee maintains the PORVs Closed with their block valves isolated.

There is no guarantee that the block valves would be able to be opened in the harsh environment inside the containment;

'"Currently to comply with 10CFR50, Appendix R, PORV breakers52-196 and 52-224 are norma11y left open.

This is to prevent the possibi1ity of."hot shorts" simultaneously in control circuits for each PORV and its associated block valve, resulting in a LOCA."

(Supplement 1 to NUREG 0737-Response to Draft Safety Evaluation - Procedures Generation,

Package, Attachment 1, page 3.)

CPCo di~cussion:

• . The NRC concern involvin~ operating with the block valves closed, and their not being qualified for the containment environment which may follow the "event" is understandable, but it is neither new nor unreviewed.

Beca~se this same situation existed when the NRC reviewed the PRA analysis of the "event" and issued.the SER, we consider that this concern has been reviewed and found acceptable.

Since the submittal, review, and approval of the PRA, several system upgrades have occurred..

Palisades no longer operates with the circuit breakers for the block valves open.

The electrical circuitry for both the PORVs and the block valves has be~n u~graded to meet Appendix R requirements.

In addition, the electri~al portions of the installation no~ meet Environmental Qualification requirements; the PORVs and block valves have been added to the "Q Li~t" and included in the lnservice Testing Program; Technical Specifications, which apply during operation, have been proposed to assu~e PORV flow path availability; and the PORVs, block valves, and associated portions of piping have been replaced* to

• provide significantly greater flow capacity assuring that they can accomplish**

the once-through-cooling function under a much broader range of Primary Coolant System conditions.

ATTACHMENT 2 Consumers Power Company Palisades Plant Docket 50-.255

• Accomplishment of Committed A~tions July 6, 1993 5.Pages

. f

• , Page 1 Accomplishment of Conunitted Actions The following is the paragraph extract from CPCo to NRC letter of April 28, 1986 containing the commitments which are the subject of the Notice of Deviation:

.The new EOPs are structure.d such that, if in-containment vita-1 instrumentation is considered. lost then the Functional Recovery Procedure is implemented. If a decision is made to not rely on any in-containment instrumentation, the result is that l} the PORVs are opened and a77 available safety injection (SI) pumps started and SI valves open; 2) auxiliary feed flow is maximized and directed to at least one SIG; 3) containment air coolers are placed in emergency configuration and containment spray manua71y maximized; and 4) al 1 available service water and component cooling water pumps operated.

The statements made in this paragraph are accomplished, by the EOPs, as described below.

We note that the paragraph above do~~ not imply that the results would occur in the listed order, or that they would be contained within a unique step in the EOPs for immediate implementation upon diagnosis of the event.

In the following discussion, the step numbers listed are taken from the current EOPs, Revision 3.

• , Page 2 Item 1:

• "The new EOPs a~e structured such that, if in-containment vital instrumentation is considered lost then the Functional Recovery Procedure is implemented."

Procedura 1 operator response to any reactor 'trip starts with ut il' i zat ion of EOP LO, Standard Post Trip Actions. That procedure's immediate actions direct.

completion of several generalized actions needed following most trips:

verification of reactor shutdown, electric power availability, PCS inventory &

pressure control, core & PCS heat removal, containment isolation & environment, and vital auxiliaries.

If any of these checks show other than normal post trip conditions, the operator is directed to a diagnostic flow chart.

In the case of the "event", this chart would direct the operator to either EOP 6.0, Excessive Steam Demand Event~ 6r to EOP 9.0, Functional Recovery Procedure, depending on the exact condition~.

If EOP 6.0 is implemented, prior to the first operator action, the following caution note is encountered:

During degraded Conta1nment conditions, the operator should not rely on any single instrument indication due to large instrument errors.

Alternate/additional instrumentation should be used to confirm trending of PCS conditions.

EOP 6.0, Step 1 then directs the operators to perform Safety Function Status Checks at'approximately 15 minute interVals.

The Safety Function Status Check sheet contains several items which, depending on the timing of the verification and the severity of the break, might not meet their acceptance criterton.

If any of the instrumentation used to confirm the safety functions should be determined to be unreliable (follow~ng the above caution) that safety function could not be considered. confirmed.

In addition, Safety Fuhction Status Check 7, Containment Isolation, *requires a verification of MSIV closure which, by the nature of the "event", could not be met.

If any safety function does not meet its acceptance criteria, EOP 6.0 Step j

• directs the operator to either retu~n to the EOP 1.0 diagnostic flow chart to re-evaluate the symptoms, or to implement EOP 9.0. Since the only proper

• choices of the EOP 1.0 diagnostic flow chart for the event" would be EOP 6.0. or 9.0, the result would be implementation of EOP 9.0, as stated.

, Page 3 Item 2:

"if a decision is made to not relyon any in-containment instr~mentation, the result is. that 1} the PORVs are opened When EOP 9.0 is implemented, prior to the first operator action, the following caution note is encountered:

During degraded Containment conditions, the operator should not rely on any single instrument indication due to* large instrument errors.

A1ternate/addifiona1 instrumentation should be used to confirm trending of PCS conditions.

EOP 9.0, step 9 directs the operator to identify the success paths currently in use for each safety function by use of the Resource Assessment trees. Resource Assessment Tree "E" directs use of either Success Path HR-4, -Once-through-cooling, or HR~3, S/G heat sink and SI pump operation, depending on exact plant conditions.

If HR-3 is selected, as it might be for less severe "events", step 22 (a step*

which is not performed se~uentially, but rather performed at any time when its "IF - THEN structure is satisfied} directs the operator to implement HR-4 if the steam generators are not capable of providing the necessary heat removal function.

Clearly, if in-containment instrumentation had been determined to be unreliable, the operator could not.conclude that the steam generators were available.

. When HR-4 is e~tered, the second step entountered directs the operator to open th~ PORVs.

The result.is that the PORVs w6uld be opened as stated.

, Page 4 Item 3:

"If a decision.is ~ade to not rely on any in-containment instrumentation, the result is that 1)... all available safety injection (SI) pumps started and SI valves open;

._.. 3) containment air coolers are placed in emergency configuration and containment spray manually maximized; and 4) all available service water and

• compon~nt cooling water pumps operated."

If PCS pressure drops below 1605 psia; as it would for the "event", the

-following actions, among others, are automatically initiated by the Safety Injection Signal (SIS):

Starting of ~11 available safety injection (SI) pumps and opening of all SI

valves, Switching of the containment air coolers to emergency configuration, and Starting of all available service water and component cooling water pumps.

If containment pressure exceeds 3.7 psi a, as it.would for the "event, the following actions, among others, are automatically initiated* by t~e Containment High Pressure (CHP) signal:

Initiation of SIS, and Startirig of all available containment spray pumps and full opening of both containment spray valves, If the SI~ or CHP setpoint is reached, or if unreliable instrumentation prevented confirmation t~at PCS pressure was above 1605 psi a, SIS a~d CHP initiation would be verified by EOP 1.0, Steps 7 and 10, respectively.

If EOP 6.-0. was implemented, initiation of SIS and CHP, and actuation of the required equipment is accomplished by Steps 5 and 12.

When EOP 9.0 is implemented, the SIS verification actions would be accomplished by Step 1 of both HR--3 and HR-4, and the remaining CHP verifi~ations by steps 3 and 4 of CA-3.

The result is that this equipment would be operated as itated.

• , Page 5 Item 4:

"If a decision is made to not rely on any in-containment instrumentation, the result is that... 2) auxiliary fe~d flow is maximized and directed to at least one S/G;"

The.starting. of one auxiliary feedwater (AFW) pump and openi_ng of the associated AFW flow control valves for eath steam generator (S/G) is accomplished automatically by the Auxiliary Feedwater Actuation Signal (AFAS) if either S/G narrow range level drops below the AFAS setpoint.

If S/G level is below th~ AFAS setpoint, as both would be in ~ase of t~e "event", or if unreliable instru~entation prevented confirmation that S/G level is above the AFAS setpoint, AFW flow of at least 165 gpm to at least one S/G (with a footnote stating that as much as 310 gpm may be necessary to recover S/G level), is vertfied by EOP 1.0, Step 9.

• If EOP 6.0 was implemented, operation

-of AFW is-accomplished by Step 17.c.

When EOP 9.0 is implemented, the same actions would be accomplished by HR-3 step 13.c (note that HR-4 is typically used when AFW is not available). While the EOPs do contain direction to isolate the most affected S/G, they do not direct the operator to terminate AFW flow.

Therefore AFW flow should be continuously supplied to at least one S/G until other means of decay heat removal, such as shutdown cooling, have been es~ablished. The result.is that this equipment would be operated as stated.

ATTACHMENT 3.

Consumers Power Company Palisades Plant Docket 50-255.

Discussion of Open Items July 6, 1993 4 Pages

.., Page 1

• Open Items Open Item 93010-01 Commitment for as found MSIV testing:

Prior to this inspection, the licensee had documented c)osure of the MSIVs on a shutdown check list, but no measurement of the closure time was taken.

The licensee has since committed to include as-found stroke tim*e testing of the MSIVs in the operational procedures.

The results of this as-found testing wi77 indicate if there was any degradation in the ability of the MSIVs to perform their function over the previous operational cycle.

The commitment to include as-found stroke time testing of the MSIVs in the operation procedures will be tracked as an open item (255/93010-0l(DRS)).

CPCo Response:

Operating Procedure GOP 9 has been revised to specify that the subject testing be performed when the MSIVs are tlo~ed during a plant cooldown.

This testing of the MSIVs was performed during the June 6, 1993 shutdown for the current refueling outage.

The valve closure times were within the Technical Specification limit.

, Page 2

• Open Item 93010-03, Procedure changes and training tin loss of instru~entat~on The licensee met their commitment in this case, but failed to fu77y address the concerns expressed by the NRC.

In particular, adequate procedure actions to cope with the loss of information or misinformation caused by the loss of instrumentation were not established.. The inspectd~s reviewed the training records related tQ this event, and found no records of training in loss of instrumentation.

Some guidance is. given in EOP 6.0 for narrow range steam generator level and pressurizer level, but this would only be applicable for a single uncomplicated steam line break (no other events in

• progress concurrent with the steam line break).

Operations personnel stated that they did not know what instrumentation would be affected, and they would have to wait for the engineering staff to

• evaluate containment conditions and determine which instrument action would

• be available for Use.

This evaluation* would take place during the event.*

The licensee failed to fully address the concerns raised by the NRC in re$ponse to item 2.

Specifica77y, adequate changes to the procedures and operator training on loss of instrumentation related to this event were not developed.

This is considered an open item (255/93010-03(DRS. CPCo Response: Palisades will provide operator training which will explain the ~event" and its possible consequences. The training will discuss use of alternate instrumentation. Analyses have been run, for the Loss of All Feedwater event, to determine if AFW or Once-Through-Cooling, initiated after S/G dryout can provide satisfactory.

• decay heat removal.

The results indicate that successful removal of decay heat can be accomplished by eithet Once-Through-Cooling using a single PORV, two charging pumps, and a single High Pressure Safety Injection {HPSI)/Containment Spray pump train, or Auxiliary Feedwater flow using a.single pump, even if initiated after steam generate~ dryout, with the primary coolant temperature approaching 545°F. The double steam generator blowdown e~ent initially cools the PCS, providing time during the PCS reheating for initiatio~ of either . cooling method. The availability of the additional train of equipment for each

• cooling method which would be available following the subject event lengthens the time available for initiation of cooling.

Pa 1 is ad es is currently running pre 1 i mi nary ana 1 yses to estab 1 i sh the containment

• and PCS responie to the event" with the existing plant configuration. These analyses will be completed to support the above mentioned operator ~raining.

Consumers Power Company considers that, unless these analyses yield unexpected results, the training discussed above combined with the existing EOPs will satisfy the NRC concern mentioned in the SER. \\. , Page 3 Op~n Item 93010-04. Suggestions f6r specific improvements to the EOPs:. The caution prior to step 12 in EOP 9.0, Success Path 9.0, states, "If maintaining an SG heat sink is immediately needed to protect the Public Health/Safety and both SGs are required to be isolated, then the Shift Supervisor may direct departing from the SG isolation steps (reference 10CFR50.54X)." The basis for this caution states "If both SGs are affected, isolating both SGs will result in a violation of HR-3 safety function acceptance criteria. Only HR-4, "Once-Through-Cooling," would be available". This caution should direct the operators to the proper procedure, HR-4. The caution, stating that the Shift Supervisor may deviate from the procedure via 10CFR50.54x, provides no useful information. In this case, actions that can provide adequate protection are not specified even though they are available. In addition, 10 CFR 50.54x is invoked in an emergency when reasonable actions that depart from a license condition or a technical specification are needed to protect the public

• health and safety and no actions consistent with licerise equivalent protection are immediately apparent.

In this case, appropriate actions can. be taken through HR-4. EOP 6.0, Attachment 2, contains graphs to account for errors in pressurizer and SG nar~ow range level. This attachment in [sic] not referred to in the. body of the procedure.. It is referred to in footnotes in the Safety Status, but no guidance is given as to when these graphs are required. This was brought to the licensee's attention with no immediate response. This is considered an open item (255/93010 (DRS)). CPCo Response: . Consumers Power company agrees that the note referencing 10 CFR 50.54x is

• unnecessary and inappropriate.

These notes wi 11 be removed during the current EOP revision effort. EOP 6.0, Atiachment 2 is referred to in several places within the EOPs, Step 6.a. (1) of Attachment* 1 to EOP 6.0 is typical. That step directs the operator to verify that S/G level is within the desired band. An asterisk following the specified level band ~ignifies an associ~ted footnota. That footnote:* "SIG level instrument decalibration can occur due to Containment atmosphere/System pressure (Refer to Attachments2 and 4)." The graphs themselves contain instruction for applying the contained information. No change is.considered necessary. , Page 4 Open-Item 93010-05, Simulator modeling improvements: The fo77owing parameters did not appear to be properly modeled on the Palisades Plant specific simulator. Containment Pres'sure - During the first scenario with a77 ECCS equipment available, the containment pressure peaked at 12 psfg. During the second scenario with a loss of off site power and only one train of ECCS equipment available, maximum containment pressure was 22 psig. Previous analyses indicated that these pressures are lower than what would be present during an actual event. Containment Temperature - The maximum containment temperature identified during either scenario was 150°F. Previous analyses

• indicated that this temperature is lower than what.would be present during an actual event.

As a result of the containment temperature and pressure discrepancies, it is believed that the training on this event had a negative effect. The oper~tors were not presented with realistic values to enable them to make the required decisio.ns. This tends to lead to a false sense* of security .and reliance on pr6cedures that do not appear adequate to recdver from this event. This is considered an open* item. (255/93010-05(DRS)). CPCo Response: Consumers Power C6mpany is currently reviewing the differences betwee~ the analytical and simulator run results for the "event". One difference that has been found is that the simulator's maximum stea~ line break {a severity level of 1.0, which was demonstrated for the NRC during the subject inspection) modeled a flow of 5.1E6 lb/hr versus the current containment analysis modeling ~t 12.9E6 Tb/hr.* The containment analysis modeling with the former S/Gs, at the time th~ PRA was completed, used an initial flow of 39.2E6 lb/hr. The simulator modeling has been increased to more closely agree with the current analyses. The review

• is continuing.. We will determine the reasons for the differences between the-simulator presentation*and analytical modeling for this event, and correct those differences which are inappropriate. There are several reasons, however, why the simulator should not duplicate the results presented FSAR or in the PRA for the "event."

The simulator model uses* "best engineering estimate" modeling versus the very* conservative "worst case estimate" {such as assuming all in-containment instrumentation fails) that licensing base analyses typically use. The initial conditions of a simulator run do not model every parameter at its limiting condition, as do the typical safety_ analyses. Rather, the simulator simply. displays the results calculated forthe conditions existing {on the simulator). at the start of the run. The simulator modeling more closely represents the plant as it is today than does the 1986 PRA analyses. In 1989 the Palisades steam generators were replaced. The new iteam generators incorporate integral flow restrictors.in the steam outlets. The containment analyses associated with the Mai~ Steam Line Break accident show that the initial rate of steam flow to the containment is reduced by a factor of 3 with the new S/Gs {12.9E6 lb/hr vs 39.2E6 lb/hr).* This change alone would slow the conta~nment pressure rise and result in a slower

• transient.

More of the steam release would occur during the period when the containment sprays are in service.. ~

ATTACHMENT 4 Consumers Power Company Palisades Plant Docket 50-255 Discussion ~f Related Commitments July 6, 1993 I Page , Page 1 Related Commitments During the course of our review or the do~uments associated with the subject commitments, four additional items, where our actions did not precisely match our commitments, were noted.

1.

In the first paragraph on page 3 of our April 28, 1986 letter, we state that caution notes will be added to the beginning of operat6r actions and Safety Function Status checks of the Loss of Coolant Accident (LOCA) and Excessive Steam Demand Event (ESDE) procedures discussing the ~ossible

• effects of degraded containment conditiOns on instrumentation. These notes
• were placed in the operator actions section of those procedures, but not in the Safety Function Status check sections. Since the notes were the first item encountered in the operator action section of the procedure~,

occurring immediately above the step which directs performance of the Safety Function Status Check, it was determined that there was no nece~sity to repeat it in the Safety Function Status check.

2.

In the third paragraph on that same page, we st~te that the upper limit of design containment temperature and pressure will be included in the LOCA. ~nd ESDE Safety Function St~t~s checksi The current revision of these EOPs contains the pressure limit, but not the temperature limit. The temperat~re limit was subsequently considered. iriappropriate for inclusion because the containment temperature instrumentation was not qualified for the post-LOCA conta~nment environment.

3.

In the third paragraph of the following page, we state that a caution will be inserted into the Steam Generator Tube Rupture (SGTR) procedure to advise the operators that steaming a S/G with a ruptured tube will cause a. loss of PCS inventory. The note does not appear in the current revision of EOP 5.0 (SG°TR). The caution was determined to be obvious and therefore unnecessary. The direction for monitoring of SIRWT lev.el, mentioned in the same ~aragraph, are found in steps 28 and 29 of EOP 5.0.

4.

On page 4 of our August 1, 1986 letter, we state that an attachment _listing* instruments outside the containment which can be used to confirm or back up instruments inside the containment will be added to the EOPs for LOCA, ESDE, and Functional Recovery. That attachment was included in the functional Recovery procedure instead; When the procedures ~ere being prepared, it was realized that the conditions which.would necessitate the use of the attachment would result in EOP 9.0 being implemented~ Therefore, the attachments would not be needed when following the LOCA or ESDE procedure. The failure to implement commitments such as these exactly a~ ~ritten or to notify the NRC of the changes, typically occur when the person writing or revising the procedure was*unaware that the specific item was the subject of a commitment. Palisades now has a commitment tracking system which is intended to prevent these kinds of errors. }}